Boiling water reactors (BWR) are featured as the plant thermal power generator within a substantial number of nuclear power facilities. These reactors are characterized in having a steam generating architecture wherein the primary coolant, water, serves both as a reaction moderator and, being brought to boiling along the reactor core, further functions as the system working fluid. Typically, this water working fluid is recirculated through the reactor core by dual, independent recirculation loops driven, in turn, by relatively large recirculation pumps.
Within the vertically oriented reactor vessel itself, the water is circulated upwardly and undergoes steam conversion through the core situated at the vessel lower region. The water-steam mixture, then typically rises through a steam separator assembly and steam dryer structure, whereupon the steam exits to a turbine drive function. The steam-water mixture within the upper region of the reactor vessel is somewhat variable in extent, its uppermost profile being referred to as the "froth level". This is, however, not easy to measure. More commonly, a water level equivalent to the pressure differential between the top of the reactor vessel and a suitable lower point in the vessel is measured. This is referred to as the "collapsed" or "induced" level of water.
An accurate and continuous monitoring of the collapsed water level within a reactor is, of course, quite important from a general performance and safety standpoint. However, investigators have found that the power output performance characteristics, particularly of smaller reactor installations, can be improved by exerting a fine control over the induced water level within the reactor vessel. See in this regard:
Termaat. K. P.; TRICOTH-III; "A Hybrid Digital/Analog Thermocouple Sensor to Measure Water Level in Nuclear Reactors.", MSC Thesis, Measurement and Control Section E. R., Eindhoven University of Technology and Division R&D of KEMA at Arnhem, The Netherlands; February, 1990. PA1 "Bicoth Water Level Sensor and Its Applicability to Measurement of In-Core Thermal/Hydraulic Performance" by Ara, et al., Proceedings of a Specialist Meeting on Core Instrumentation and Reactor Assessment, Organization for Economic Cooperation and Development, Fredikstad, Norway, 1983. PA1 providing a sequence of electrically, serially coupled, junction defining thermocouple element pairs of successively increasing lengthwise extent; PA1 foldably configuring the thermocouple element pairs in a parallel, mutually adjacent vertical orientation wherein alternate junctions commencing with the first junction of the sequence are located at a common predetermined base level within the vessel, and wherein alternate junctions commencing with the second junction are each located at a predetermined unique level within the vessel spaced from the base level; PA1 positioning a heating component within the vessel adjacent to and in thermal exchange relationship with the thermocouple element pairs; PA1 applying heat to the thermocouple element pairs from the heating component through and under the heat dissipating influence of the steam environment above the water level and under the heat dissipating influence of the water environment below the water level; PA1 detecting the amplitude of voltage output of the sequence of thermocouple element pairs; and PA1 correlating the detected amplitude with a calibration value to derive the position of the level of water within the vessel. PA1 providing a substantially linear sequence of electrically serially coupled, junction defining thermocouple element pairs from first to last; PA1 positioning the linear sequence of thermocouple element pairs vertically within the vessel extending from within the steam environment into the water environment; PA1 providing a heating assembly formed as a linear sequence of mutually spaced, discrete heating components from first to last; PA1 positioning the heating assembly within the vessel such that each discrete heating component, from first to last, is located in thermal exchanging adjacency with a select portion of respective first to last thermocouple element pairs; PA1 applying heat to the thermocouple element pairs from first to last from respective first to last heating components through and under the heat dissipating influence of the steam environment above the water level and under the heat dissipating influence of the water environment below the water level to establish a differential of temperature across each thermocouple element pair; PA1 detecting the amplitude of the output signal of the linear sequence of thermocouple element pairs; and PA1 correlating the detected amplitude with a calibration value to derive the position of the level of water within the vessel.
A variety of techniques for monitoring or gauging the collapsed level of water within a reactor have been proposed and evaluated by investigators. For example, differential pressure transducers have been employed on the outside of the reactor vessel utilizing measuring standpipes with damping restrictions at the lower ends and water filled reference standpipes with small condensation spheres. Additionally, ultrasonic waves have been investigated as well as capacitive sensors. More recently, the utilization of heated thermocouples within the reactor vessel itself have been the subject of investigation. For example, a digital form of water level evaluation may be developed utilizing a sequence of thermocouples having selectively positioned junctions and configurations to develop a Gray type code output representing water level. A temperature differential is experienced across select ones of these thermocouple structures by virtue of their thermal influence with an elongate heated wire located adjacent to them. For example, at a water-steam interface, the heated wire will experience a greater heat sinking effect from the water component or environment than the steam component or environment to evoke a temperature differential and corresponding voltage signal output. Generally, a temperature transition function is witnessed at the water level across a given involved thermocouple which has been employed for extrapolation purposes to achieve even more accurate water level monitoring. The binary coded form of readout from such an arrangement has been found effective; however, a substantial number of leads may be involved in developing the monitoring structure which, necessarily, must be accommodated for by the supporting and penetrating structure leading through the reactor vessel top component. See additionally, the following publication: